1. Field of the Invention
The present invention relates generally to fuel assemblies for a nuclear reactor and, more particularly, is concerned with a boiling water reactor (BWR) fuel assembly employing improvements in spacer and fuel bundle design for enhanced thermal-hydraulic performance.
2. Description of the Prior Art
Typically, large amounts of energy are released through nuclear fission in a nuclear reactor with the energy being dissipated as heat in the elongated fuel elements or rods of the reactor. The heat is commonly removed by passing a coolant in heat exchange relation to the fuel rods so that the heat can be extracted from the coolant to perform useful work.
In nuclear reactors generally, a plurality of the fuel rods are grouped together to form a fuel assembly. A number of such fuel assemblies are typically arranged in a matrix to form a nuclear reactor core capable of a self-sustained, nuclear fission reaction. The core is submersed in a flowing liquid, such as light water, that serves as the coolant for removing heat from the fuel rods and as a neutron moderator. Specifically, in a BWR the fuel assemblies are typically grouped in clusters of four with one control rod associated with each four assemblies. The control rod is insertable within the fuel assemblies for controlling the reactivity of the core. Each such cluster of four fuel assemblies surrounding a control rod is commonly referred to as a fuel cell of the reactor core.
A typical BWR fuel assembly in the cluster is ordinarily formed by a N by N array of the elongated fuel rods. The bundle of fuel rods are supported in laterally spaced-apart relation and encircled by an outer tubular channel having a generally rectangular cross-section. The outer flow channel extends along substantially the entire length of the fuel assembly and interconnects a top nozzle with a bottom nozzle. The bottom nozzle fits into the reactor core support plate and serves as an inlet for coolant flow into the outer channel of the fuel assembly. Coolant enters through the bottom nozzle and thereafter flows along the fuel rods removing energy from their heated surfaces.
In a fuel assembly of this type the fuel rods in the central region of the bundle thereof may be undermoderated and overenriched. In order to remedy this condition by increasing the flow of moderator water through this region of the assembly, an elongated centrally-disposed water cross is frequently used in the assembly, as disclosed in the above cross-referenced Barry et al, Doshi and Lui patent applications. The central water cross has a plurality of four radial panels which together form a cruciform water flow channel which divides the fuel assembly into four, separated elongated compartments, with the bundle of fuel rods being divided into mini-bundles disposed in the respective compartments. The water cross thus provides a centrally-disposed cross-shaped path for the flow of subcooled neutron moderator water within the channel along the lengths of, but separated from, adjacent fuel rods in the mini-bundles thereof. The fuel rods of each mini-bundle extend in laterally spaced apart relationship between an upper tie plate and a lower tie plate and connected together with the tie plates to comprise a separate fuel rod subassembly within each of the compartments of the channel. The water cross has approximately the same axial length as the fuel rod subassemblies, extending between the upper and lower tie plates thereof.
Additionally, a plurality of grids or spacers, for example six in number, are disposed at axially displaced positions along the fuel rods of each fuel rod subassembly to maintain the fuel rods in their laterally spaced relationships. The spacers are introduced to maintain the desired fuel bundle configuration and to prevent excessive fuel rod bow and flow induced vibrations. Thus, the spacers provide significant benefits from a structural standpoint.
However, the spacers are undesirable from a thermal-hydraulic performance standpoint. Specifically, they cause increased pressure drop (consequently higher coolant pumping powers are needed) and reduce the bundle critical power ratio (CPR) due to turbulence generated. The critical heat flux (CHF) deteriorates because the flow turbulence, induced by the spacers, strips away the liquid film from the heated fuel rod. This causes the bundle heat transferred to the coolant to drop significantly. Consequently heat builds up in the fuel rod causing it to overheat, and in some cases "burnout", releasing radioactivity into the coolant. This imposes a safety constraint. In order to avoid this possibility (i.e., overheating), safety analysis calculations are performed for each cycle to demonstrate that adequate margin to the CPR is always available, even under the worst postulated transient. Dedicated computer hardware performs continuous on-line surveillance of the fuel thermal margins to insure that adequate CPR margin is always maintained during reactor operations.
It is well known that different spacers exhibit variable CHF performance. Further, this performance varies radially with the fuel rod location, and in general correlates with the spacer pressure drop. The smaller the value of the pressure drop, the better is the CHF performance. However, the spacer pressure drop characteristics vary radially within the fuel rod bundle. Due to this, spacers can either be so-called "corner", "side" or "interior" limiting. If a rod peaking occurs in the limiting location, CHF occurs much earlier, and thus causes deterioration in CPR margins. It should also be noted that not all spacers for substantially identical fuel rod array configurations (having the same N.times.N array of fuel rods) will behave identically.
Consequently, a need exists for some design modifications to fuel assembly spacers in order to avoid or minimize the detrimental effects of CPR limiting locations within the spacers.